400~1500 MeV质子轰击铅靶和钨靶的出射中子能谱的FLUKA和Geant4模拟研究

在ADS散裂靶系统的优化设计中,蒙特卡罗方法结合可靠的散裂反应理论模型进行中子学计算具有重要的作用。本工作利用Geant4程序中的INCLXX模型、BIC模型以及BERT模型和FLUKA程序分别模拟了597 MeV和1 500 MeV质子轰击薄铅靶不同出射角度的中子双微分截面,500,1 500 MeV质子轰击厚铅靶不同出射角度的中子双微分产额,以及400,600,800,1 000和1 200 MeV质子轰击厚钨靶在反角方向（175 °）的中子双微分产额,并与实验数据进行比较。研究表明,对于薄铅靶,Geant4程序的INCLXX模型和FLUKA程序模拟结果与实验符合得更好。能量在10～40 MeV范围内,BIC模型模拟结果明显高于实验数据,而BERT模型模拟结果略微低于实验数据。对于厚铅靶,在40 MeV左右所有的模拟结果都低于实验数据。对于厚钨靶,Geant4程序的BIC模型和FLUKA程序与实验数据符合得较好,INCLXX模型在能量高于60 MeV时模拟结果低于实验数据,BERT模型与实验数据差异较大。总体来看,Geant4程序的INCLXX模型和FLUKA程序进行ADS散裂靶相关的中子学的计算是合理和可靠的。The reliable Monte Carlo simulation codes coupled with nuclear reaction models play an important role in the neutronic calculation for the design and optimization of the ADS spallation target. In this work, the double differential cross sections at different angles produced from a thin lead target bombarded with 597 and 1 500 MeV protons, the neutron energy spectra at different angles produced from a thick lead target bombarded with 500 and 1 500 MeV protons, and the neutron energy spectra in the backward direction(175°) produced from a thick tungsten target bombarded with 400, 600, 800, 1 000 and 1 200 MeV protons are calculated with the Geant4 code coupled INCLXX, BIC and BERT models and the FLUKA code. The calculations are compared with the available experimental data. The results show that, for the thin lead target, the calculations with the Geant4 coupled INCLXX model and FLUKA code well reproduce the experimental results. In a energy range from 10 to 40 MeV, BIC model obviously overestimates the experimental results, and BERT model slightly underestimates the experimental results. For the thick lead target, all of the calculations underestimate the experimental results around 40MeV. For the thick tungsten target, the Geant4 coupled BIC model and FLUKA code well reproduce the experimental results. INCLXX model underestimates the experimental results above 60 MeV. BERT model bad reproduces the experimental results. Overall, the neutronic calculations with the Geant4 code coupled INCLXX model and FLUKA code for the ADS spallation target is reasonable and reliable.     

HIRFL-CSR系统中RIBLLⅡ上中子墙设计的模拟计算

HIRFL-CSR系统中RIBLLⅡ外靶实验终端上的中子探测装置是用来开展放射性束物理以及利用放射性束开展非对称核物质性质等研究的一个重要的探测装置. 介绍了在RIBLLⅡ上计划建造的中子墙的探测原理以及基于Geant4对其所进行的蒙特卡罗模拟, 给出了中子墙的设计方案、优化的设计参数以及中子探测的性能指标.     

利用GEANT4研究轻带电粒子诱发反应出射中子双微分产额

轻带电粒子诱发反应产生次级中子的研究对于加速器屏蔽设计和优化具有重要意义。利用Geant4程序结合INCL、BIC、BERT物理模型分别计算了33 MeV的d核、65 MeV的3He核和4He核轰击厚的碳、铜和铅靶在轻带电粒子诱发反应产生次级中子的研究对于加速器屏蔽设计和优化具有重要意义。利用Geant4程序结合INCL、BIC、BERT物理模型分别计算了33 MeV的d核、65 MeV的3He核和4He核轰击厚的碳、铜和铅靶在$0^{\circ}$,$15^{\circ}$,$45^{\circ}$,$75^{\circ}$和$135^{\circ}$等方向出射中子的双微分产额,并与实验数据进行了比较。研究表明,对于33 MeV的d核诱发的核反应,INCL模型的计算结果基本上再现了碳靶和铜靶的实验数据,但高估了铅靶直接过程产生的中子。BIC模型和BERT模型的计算结果没有重现弹核削裂过程对应的宽峰。对于65 MeV的3He核诱发的核反应,三个模型的计算结果均未能重现前向角弹核削裂过程产生的中子,但在$15^{\circ}$,$45^{\circ}$,$75^{\circ}$和$135^{\circ}$上三个模型的计算结果与实验数据符合较好。对于65 MeV的4He核诱发的核反应,INCL模型的计算结果与碳靶和铜靶的实验数据符合较好,但低估了铅靶的中子产额。BIC模型和BERT模型的计算结果低估了碳靶的实验数据,且在大角度上略微高估了铅靶的实验数据。     

用于神光Ⅲ燃料面密度诊断的下散射中子探测系统设计

根据辐射屏蔽后5.5 m测点处的辐射场情况,分别设计了电流型探测器系统和成像型探测器系统。通过Geant4数值模拟分析可得：在面密度达到10 mg/cm2、初级中子产额为1012时,电流型探测器系统满足测量的信噪比,信噪比达到40∶1;在面密度达到10 mg/cm2、初级中子产额为1011时,成像型探测器系统满足测量的信噪比,信噪比好于10∶1;面密度增大时,信噪比有所改善;但是当初级中子产额达到1012时,出现中子信号重叠现象,可通过缩短曝光时间或者减小塑料闪烁体厚度来降低中子重叠率。     

用于神光Ⅲ燃料面密度诊断的下散射中子探测系统设计

根据辐射屏蔽后5.5m测点处的辐射场情况,分别设计了电流型探测器系统和成像型探测器系统。通过Geant4数值模拟分析可得:在面密度达到10mg/cm2、初级中子产额为1012时,电流型探测器系统满足测量的信噪比,信噪比达到40∶1;在面密度达到10mg/cm2、初级中子产额为1011时,成像型探测器系统满足测量的信噪比,信噪比好于10∶1;面密度增大时,信噪比有所改善;但是当初级中子产额达到1012时,出现中子信号重叠现象,可通过缩短曝光时间或者减小塑料闪烁体厚度来降低中子重叠率。     

中子针孔成像点扩展函数模拟研究

应用Geant4系统对中子针孔成像的点扩展函数进行了模拟研究.建立了点扩展函数的数学模型,并将其分布图像与能量沉积数据的分布图像作了比较,可看出点扩展函数与能量沉积数据吻合得较好.研究结果表明,在一定偏移量范围内,拟合得到的点扩展函数更好,能得到较好的图像分辨率.通过对不同入射能量在一定偏移量入射下的点扩展函数讨论认为,入射能量增大虽然会增强图像的对比度,但会降低分辨率.     

激光量热法测量K9基片的表面吸收和体吸收

表面抛光可能给K9基片带来额外的杂质和吸收,分离K9基片的表面吸收率与体吸收率有助于改进基片的加工质量和抛光工艺,对抗损伤能力研究具有重要意义。分析了激光量热法测量弱吸收的原理,采用符合ISO 11551要求的激光量热计测量K9基片的弱吸收。对相同工艺抛光的不同厚度K9基片进行了弱吸收表征,实验发现K9基片的弱吸收随着厚度增加近似线性增大。推导了表面吸收率和体吸收率的计算式,实验得出本样品的表面吸收率为1.2110-5,体吸收率远大于表面吸收率,体吸收系数为1.7210-3/cm。实验结果显示所用K9样品的吸收主要来自于材料本身,改善抛光工艺对降低其吸收率作用不大。     

B4C_Al材料中子吸收性能检测设备研制

当前基于燃耗信任制的乏燃料密集贮存方式,对乏燃料水池格架中子吸收材料的可靠性和有效性,都提出了更高的要求。在格架材料生产和使用过程中需要对其中子吸收性能(硼含量)进行无损检测和监测,针对这两个方面的需求,我们研制了核电厂乏燃料水池格架B₄C_Al中子吸收材料检测设备。该检测设备主要由中子源(3枚252Cf放射源)、中子探测器(10个锂玻璃组成的探测阵列)、中子屏蔽准直和慢化系统等组成,通过测量中子透射率来推算待测样板上各个测量点的10B面密度,从而达到对于乏燃料水池贮存格架材料B₄C_Al合金硼含量的无损检测。使用该套设备进行了两种B₄C_Al合金20 cm×30 cm悬挂样片的检测,结果可靠。该B₄C_Al材料中子吸收性能检测设备为国内首创,推动了我国含硼中子吸收材料的无损检测研究,能为核电厂乏燃料水池的临界安全监测提供有力保障。     

Geant4在CR39叠层探测器中的应用

利用Geant4程序计算了多种不同组合CR39叠层探测器的中子能量响应函数,研究结果表明,中子能量响应函数曲线的半高宽和探测下限分别由聚乙烯的厚度和吸收片的厚度决定,峰值由聚乙烯和吸收片的厚度共同决定。利用加速器准单能氘氘中子源进行辐照实验,获得了4种不同叠层组合探测器的2.45MeV中子探测效率,并与计算曲线进行了比较,结果吻合较好,验证了Geant4程序计算结果的可靠性。     

介质型快中子探测器中子直照响应模拟研究

 介质型快中子探测器的中子直照响应对其设计和实际应用具有重要影响。利用Geant4编制Monte-Carlo模拟程序对此型探测器的中子直照响应进行了计算。模拟了探测器芯子常用材料在中子入射时带电重粒子的发射情况,发现电荷收集极、卡阈吸收片和后高压绝缘层材料的选择对中子直照响应特性有直接影响。在现有的以聚四氟乙烯为卡阈吸收片和后高压绝缘层材料的探测芯子中,直照响应给总的中子灵敏度带来约6%的贡献。提出了一种使用高电阻率石墨作为卡阈吸收片和后高压绝缘层材料的探测芯子的优化结构,可有效降低中子直照响应。     

基于Geant4的反应堆临界及动态计算方法

Geant4是一款基于C++面向对象技术的蒙特卡罗开发程序包,可以模拟各种已知粒子与物质之间的相互作用。然而该程序包没有提供临界源功能,无法直接用于反应堆物理计算。因此,利用Geant4提供的基础物理模型和粒子跟踪控制等功能,用两种不同方法实现了临界源的设置,实现了基于Geant4的反应堆静态计算程序G4-RSM和反应堆动态计算程序G4-RDM。两个程序均可用于反应堆临界计算,与MCNP计算结果相对误差在5%以内。G4-RDM程序除可用于临界计算外,还可用于模拟堆内事故工况下的中子学瞬态变化。     

质子照相中基于能量损失的密度重建

提出了一种质子能量在中高能时利用能量损失进行密度重建的方法,并利用Bethe-Bolch公式给出了利用能量损失进行密度重建的方程及条件.针对1.6 GeV的质子能量,通过定量计算常见材料的阻止本领,得出质子能量在1.45–1.6 GeV范围内时,材料的阻止本领的变化率小于1%,可近似为常数.最后,通过理论计算和Geant 4模拟,得出质子能量在1.6 GeV时,可以对面密度为113 g/cm²的缩比法国实验客体进行密度重建.     

Monte Carlo simulation of the property of a scintillation bar in the multi-neutron correlation spectrometer

To perform a kinematically complete measurement of the dissociation reaction for neutron-rich nuclei, a multi-neutron correlation spectrometer is proposed at Peking University. A Monte Carlo simulation code based on GEANT4 is developed for a single scintillation bar which processes not only the energy deposition but also the light propagation in the scintillator and the light collection and conversion to signal at the end of the bar in a realistic way. The simulating method is described in detail in this paper, and the timing and position resolutions and detector efficiency are studied based on the simulation and compared with the experimental results. A new method of crosstalk rejection has been demonstrated to be important for the design of the whole spectrometer.     

Monte Carlo simulation of the property of a scintillation bar in the multi-neutron correlation spectrometer

To perform a kinematically complete measurement of the dissociation reaction for neutron-rich nuclei, a multi-neutron correlation spectrometer is proposed at Peking University.A Monte Carlo simulation code based on GEANT4 is developed for a single scintillation bar which processes not only the energy deposition but also the light propagation in the scintillator and the light collection and conversion to signal at the end of the bar in a realistic way. The simulating method is described in detail in this paper, and the timing and position resolutions and detector efficiency are studied based on the simulation and compared with the experimental results.A new method of crosstalk rejection has been demonstrated to be important for the design of the whole spectrometer.     

Study of neutron response for two hybrid RPC setups using the GEANT4 MC simulation approach

The present article describes a detailed neutron simulation study in the energy range 10~(-10) MeV to 1.0 GeV for two different RPC configurations.The simulation studies were taken by using the GEANT4 MC code.Aluminum was utilized on the GND and readout strips for the (a) Bakelite-based and (b) glass-based RPCs.For the former type of RPC setup the neutron sensitivity for the isotropic source was S_n=2.702×10~(-2) at E_n=1.0 GeV, while for the latter type of RPC, the neutron sensitivity for the same source was evaluated as S_n=4.049×10~(-2) at E_n=1.0 geV.These results were further compared with the previous RPC configuration in which copper was used for ground and pickup pads.Additionally A1 was employed at (GND+strips) of the phosphate glass RPC setup and compared with the copper-based phosphate glass RPC.Good agreement with sensitivity values was obtained with the current and previous simulation results.     

Study of neutron response for two hybrid RPC setups using the GEANT4 MC simulation approach

The present article describes a detailed neutron simulation study in the energy range 10^-10 MeV to 1.0 GeV for two different RPC configurations. The simulation studies were taken by using the GEANT4 MC code. Aluminum was utilized on the GND and readout strips for the （a） Bakelite-based and （b） glass-based RPCs. For the former type of RPC setup the neutron sensitivity for the isotropic source was Sn = 2.702 × 10^-2 at En = 1.0 GeV, while for the latter type of RPC, the neutron sensitivity for the same source was evaluated as Sn = 4.049 × 10^-2 at En = 1.0 GeV. These results were further compared with the previous RPC configuration in which copper was used for ground and pickup pads. Additionally A1 was employed at （GND＋strips） of the phosphate glass RPC setup and compared with the copper-based phosphate glass RPC. Good agreement with sensitivity values was obtained with the current and previous simulation results.     

Monte Carlo simulation of exposure factor

In this paper, we simulate the exposure factor by a simple model of a free-air ionization chamber with the Monte Carlo programme Geant4. Special emphasis is placed on the discussion of the exposure factor related to parameters of the chamber model. The reason for the variation in exposure factor with incident ray energy is also analysed in terms of reaction cross section for different types of reactions. The obtained results indicate that our simulation is accurate in the calculation of the exposure factor and can serve as a reference in designing air ionization chambers.     

Influence of thyroid volume reduction on absorbed dose in 131I therapy studied by using Geant4 Monte Carlo simulation

A simulation study has been performed to quantify the effect of volume reduction on the thyroid absorbed dose per decay and to investigate the variation of energy deposition per decay due to β- and γ-activity of 131I with volume/mass of thyroid, for water, ICRP- and ICRU-soft tissue taken as thyroid material. A Monte Carlo model of the thyroid, in the Geant4 radiation transport simulation toolkit was constructed to compute the β- and γ-absorbed dose in the simulated thyroid phantom for various values of its volume. The effect of the size and shape of the thyroid on energy deposition per decay has also been studied by using spherical, ellipsoidal and cylindrical models for the thyroid and varying its volume in 1-25 cm3 range. The relative differences of Geant4 results for different models with each other and MCNP results lie well below 1.870%. The maximum relative difference among the Geant4 estimated results for water with ICRP and ICRU soft tissues is not more than 0.225%. S-values for ellipsoidal, spherical and cylindrical thyroid models were estimated and the relative difference with published results lies within 3.095%. The absorbed fraction values for beta particles show a good agreement with published values within 2.105% deviation. The Geant4 based simulation results of absorbed fractions for gammas again show a good agreement with the corresponding MCNP and EGS4 results （±6.667%） but have 29.032% higher values than that of MIRD calculated values. Consistent with previous studies, the reduction of the thyroid volume is found to have a substantial effect on the absorbed dose. Geant4 simulations confirm dose dependence on the volume/mass of thyroid in agreement with MCNP and EGS4 computed values but are substantially different from MIRD8 data. Therefore, inclusion of size/mass dependence is indicated for 131I radiotherapy of the thyroid.     

An efficient numerical tool for dose deposition prediction applied to synchrotron medical imaging and radiation therapy

Medical imaging and radiation therapy are widely used synchrotron‐based techniques which have one thing in common: a significant dose delivery to typically biological samples. Among the ways to provide the experimenters with image guidance techniques indicating optimization strategies, Monte Carlo simulation has become the gold standard for accurately predicting radiation dose levels under specific irradiation conditions. A highly important hampering factor of this method is, however, its slow statistical convergence. A track length estimator (TLE) module has been coded and implemented for the first time in the open‐source Monte Carlo code GATE/Geant4. Results obtained with the module and the procedures used to validate them are presented. A database of energy‐absorption coefficients was also generated, which is used by the TLE calculations and is now also included in GATE/Geant4. The validation was carried out by comparing the TLE‐simulated doses with experimental data in a synchrotron radiation computed tomography experiment. The TLE technique shows good agreement versus both experimental measurements and the results of a classical Monte Carlo simulation. Compared with the latter, it is possible to reach a pre‐defined statistical uncertainty in about two to three orders of magnitude less time for complex geometries without loss of accuracy.